IR 05000244/1985015
| ML17254A564 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/25/1985 |
| From: | Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17254A563 | List: |
| References | |
| RTR-NUREG-0696, RTR-NUREG-0737, RTR-NUREG-696, RTR-NUREG-737, TASK-***, TASK-08-04, TASK-8-4, TASK-RR, TASK-TM 50-244-85-15, GL-85-12, IEB-79-06, IEB-79-6, NUDOCS 8510040209 | |
| Download: ML17254A564 (24) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-244/85-15 Docket No. 50-244 License No.
DPR-18 Priorit Cate or C
Licensee:
Rochester Gas and Electric Cor oration 49 East Avenue Rochester New York 14649 Facility Name:
R.
E. Ginna Nuclear Power Plant Inspection at:
Ontario New York Inspection Conducted:
Au ust
'1985 throu h Au ust
1985 Inspector:
W. A. Cook, Reside Inspect
, Ginna Approved by:
J
. Linvill, Chief, Rea or roject Section No.
2C, DRP dat Ins ection Summar
Ins ection on Au ust
1985 throu h Au ust
1985 Re ort No. 50-244/8S-1S Areas Ins ected:
Routine, onsite, regul and backshift inspection by the resident inspector (105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />).
Areas inspected included: plant activities during routine power operations; licensee action on previous findings; surveil-lance testing; review of service water leak inside containment; review of TMI Action Plan implementat,ion; Integrated Plant Safety Assessment Review; spent fuel cask unloading; and inspection of accessible portions of the facility during plant tours.
Results:
In the nine areas inspected, no violations were identified.
A current status of TMI Action Plan items is documented in paragraph 6.
A service water leak inside containment is discussed in paragraph 5.
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DETAILS Persons Contacted During this inspection period, the inspector interviewed and talked with operator s, technicians, engineering and supervisory level personnel.
Licensee Action on Previous Ins ection Findin s
(Closed)
Inspector Follow-up Item (84-10-03):
Technical Specification change for RCS vents.
Amendment No.
9 to Ginna Station Technical Specifications was issued by Zwolinski to Kober letter, dated July 30, 1985.
Technical Specification sections 3. 1. 1.6 and 4.3.5.6 delineate the operability and surveillance requirements, respectively, of the recently installed reactor coolant system vents.
(Closed) Violation (85-10-01):
Failure to incorporate containment temperature monitoring system in station surveillance program.
On May 29, 1985, the inspector identified that the containment temperature monitoring system used to satisfy a Technical Specification Limiting Condition for Operation was not incorporated in a station surveillance program.
The inspector reviewed the licensee's corrective action documented in Kober to Mur ley letter, dated August 5, 1985.
The inspector verified that the containment temperature monitoring system has been incorporated into Administrative Procedure, A-1105,
"Calibration and/or Test Surveillance Program for Instrumentation/Equip-ment of Safety-related Components".
(Closed)
Inspector Follow-up Item (79-BU-06):
Review of Operational errors and system misalignments identified during TMI incident.
The action items identified by these bulletins have been completed by the licensee and reviewed as acceptable, with one exception.
This exception ( see details in paragraph 7.6.3) is being tracked separately by the TMI Action Plan review.
This item is administratively closed.
(Closed)
Unresolved Item (79-08-08):
Acceptability of licensee response to IEB 79-06A for RCP operation, termination of HHSI, and implementation of actions specified in that response.
This unresolved item is discussed in detail in paragraph 6.b.3 of this report.
Final evaluation of the licensee's RCP tripping criteria is currently being conducted by NRR.
The RCP tripping criteria evaluation is being tracked by the TMI Action Plan review.
The remaining IEB 79-06A responses have been reviewed, found acceptable and determined to have been properly implemented.
This item is administratively closed.
The inspector had no further question.
Review of Plant 0 erations a.
Throughout the reporting period, the inspector reviewed routine plant operations.
The reactor has been operating at full power since June 8, 1985.
On August 7, 1985, Commissioner Lando Zech met with licensee representatives at Ginna station.
Commissioner Zech was given a
brief presentation by the licensee, toured the station and inter-viewed station operators.
A debriefing with licensee management was held at the conclusion of the visit.
On August 27, 1985, the inspector conducted a tour of the licensee's off site training facilities which will be used for INPO accreditation training of non-licensed station workers.
The inspector discussed licensee progress in meeting a December 1985 goal for INPO evaluation of training programs for accreditation of the Instrumentation
& Control technicians, Electrical 5 Mechanical maintenance personnel, and Chemistry 8 Health Physics technicians.
A recent turnover in'non-licensing accreditation training personnel may impact the December 1985 goal.
b.
During the course of the inspection, the inspector toured accessible plant areas.
Items reviewed include radiation protection controls, plant housekeeping, fire protection, equipment tagging and security.
Inspector tours of the control room this inspection period included review of shift manning, operating logs and records, and equipment and monitoring instrumentation status.
d.
Safety system valves and electrical breakers were verified to be in the position or condition required for the applicable plant mode as specified by Technical Specifications and plant lineup procedures.
This verification included routine control board indication review and conduct of a partial systems lineup check of the 1A and 1B Emergency Diesel Generators on August 6.
4.
Survei 1 1 ance Testin a.
The inspector witnessed the performance of surveillance testing of selected components to verify that the test procedure=was properly approved and adequately detailed to assure performance of a satisfactory surveillance; test instrumentation required by the procedure was calibrated and in use; the test was performed by qualified personnel; and the test results satisfied Technical Specifications and procedural acceptance criteria, or were properly resolve b.
The inspector witnessed the performance of a portion of the following test:
Reactor Plant Systems Operation Procedure, S-15. 1, "Flux Mapping Normal Procedure",
Revision No. 26, dated 3/13/85, performed on August 28, 1985.
Inspector review of the approved surveillance procedure determined that initial conditions and precautions identified in the procedure do not specify the steady-state conditions to be maintained while performing the surveillance.
Changing core parameters would invalidate the accuracy of flux mapping results.
The inspector discussed the procedural adequacy with the Reactor Engineer and he resolved to make appropriate changes to S-15. 1 to ensure reactor steady-state conditions are properly identified in the surveillance procedure and maintained during the performance of the surveillance.
5.
Service Water Leak Inside Containment Following the completion of Periodic Test (PT)-2.7, "Service Water Systems" on July 1, 1985 at 10:30 AM, the control room operator s observed a significant decrease in the Containment Vessel 'A'ump pump actuation interval.
An investigation by the licensee to determine the cause of leakage inside containment was immediately initiated.
A Reactor Coolant System inventory balance and containment radiation, dewpoint and recirculation fan cooler condensate collection systems checks indicated no change.
The leakage rate to sump 'A'as calculated to be approximately four to five gallons per minute and a service water leak became suspect.
While preparations were being made for a containment entry, a water sample from sump 'A'as taken and control room operators commenced the sequential isolation of each containment vessel recirculation fan cooler to expedite the identification of the leakage source.
At 12:03 PM, isolation of the 'A'ecirculation fan cooler system resulted in a decrease in sump 'A'ump actuations.
Personnel entering containment were informed of this determination and their further investigation found that, one of four 1/8 inch diameter threaded motor cooler drain plugs on the 'A'ecirculation fan cooler unit had corroded away and blown out under system pressure.
The displaced motor cooler drain plug was replaced and the integrity of all other recirculation fan cooler drain plugs was verified.
The
'A'ecirculating fan cooler was restored to service at 2:40 PM July 1, 1985.
Approximately 1100 gallons of service water leaked to sump
'A'nd was pumped to the Waste Holding Tan t
A similar event occurred on February 11, 1981 as reported in Licensee Event Report No.81-004.
All fan cooler motor drain plugs were replaced after this February ll, 1981 event.
The inspector determined that the 1/8 inch diameter drain plugs are made of carbon steel and serve as cathodic protection for the recirculation fan motor cooler units.
Sub-sequent to the 1981 event, annual containment recirculation fan cooler inspections only included a visual inspection of the drain plugs.
The licensee stated that future drain plug inspections will consist of a physical check of the plug integrity.
This event was properly reported to the NRC via the ENS on July 1, 1985 and a follow-up written report submitted as an enclosure to Kober to NRC letter dated July 12, 1985.
6.
Review of Three Mile Island TMI Action Plan Im lementation The inspector conducted a review of the implementation of TMI Lessons Learned to: determine the status of previously inspected items; inspect recently completed TMI Action Plan items; and to determine the status of open items and their projected completion dates.
a.
Closed TMI Action Plan Items The inspector reviewed closed TMI Action Plan items to determine if any problems involving operability, inadequate maintenance, surveillance, training or procedural compliance have surfaced since their implementation and final Inspection
& Enforcement (IE)
inspector closeout.
A review of applicable completed maintenance surveillance procedures, a review of equipment history records and inspector discussions with licensee representatives indicate there have been no significant problems identified with these completed Action Plan items.
(Attachment A lists Action Plan items in this category)
b.
Recentl Com leted TMI Action Plan Items The inspector reviewed the licensee's actions on the following items to verify that the licensee commitments were met:
1.
Containment Water Level Monitor 11.F. 1 Attachment
This action plan item required the licensee to provide continuous indication of containment water level in the Control Room.
A narrow range instrument shall cover the range from the bottom to the top of the containment sump and a wide range instrument shall cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity.
In Maier to Crutchfield letter, dated December 15, 1980, the licensee committed to install a narrow range containment water level instrument (sump A) in accordance with
Regulatory Guide
~ 89.
In addition, the existing wide range containment water level instrument (sump B), designed to monitor water depths corresponding to 500,000 gallons, was to be replaced with environmentally qualified equipment.
An inspector review of the narrow range containment water level instruments was documented in Inspection Report No.
50-244/84-01, paragraph 6.
At the time of that inspection, multiple transducer (Viatran Model 511-P10) fai lures had been experienced and the failure mechanism was believed to be water inleakage due to a poor seal between the sensor cable and transducer casing.
Viatran modified the seal arrangement and the licensee replaced the transducers with new transducers having a modified seal during the 1984 outage.
No subsequent sump A level instrument problems have been observed.
The inspector verified that continuous redundant wide range (sump B) containment water level indication is provided in the control room.
The wide range level monitoring system is composed of a series of level switches located at 8, 78, 113, 180 and 214 inches from the bottom of sump B.
The two independent sets of level switches feed separate level indication lights and recorder s in the control room.
The inspector reviewed the licensee's Design Criteria, Safety Analysis, Station Modification package and equipment qualification reports conducted by Wyle Laboratories for the supplier, Transamerica Delaval, Inc.,
and found no discrepancies.
A review of completed calibration and maintenance procedures and discussion with licensee representatives indicate that no significant problems have been experienced with the sump B level instrumentation since being upgraded in the Spring of 1982.
References:
- CP-2039A, "Calibration and/or Maintenance of LT-2039 'A'ontainment Sump Level", Revision 2, dated 2/27/8 CP-2044,
"Calibration and/or Maintenance of LT-2044 'A'ontainment Sump Level", Revision 2, dated 2/27/85.
-
EWR 3262, Design Criteria,
"Sump B Level Indication", Revision 3, dated 4/20/82.
-
EWR 3262, Safety Analysis,
"Sump B Level Indication", Revision 3, dated 4/20/8 CP-942, "Calibration and/or Maintenance of Containment Sump B ¹1 Level Switches and Indicating Lights", Revision 0, dated 1/27/8 CP-493, "Calibration and/or Maintenance of Containment Sump.B ¹2 Level Switches and Indicating Lights", Revision 0, dated 2/2/83.
- Transamerica Delaval, Inc. to RG&E Corp. letter, dated December 20, 1983.
- Wyle Laboratories, Nuclear Environmental gualification Test Reports 45700-1 and 45700-2 dated December 8, 1982 and December 14, 1982, respectivel Crutchfield to Kober letter, dated May 10, 1984.
2.
Containment H dro en Monitor II.F Attachment
This action plan item required the licensee to provide continuous indication of hydrogen concentration in the containment atmosphere after a design basis event.
The licensee committed to install a containment hydrogen monitoring system in accordance with NUREG-0737 as documented in White to Ziemann letter, dated November 19, 1979.
The inspector verified that the licensee has installed and is properly maintaining two redundant containment hydrogen concentration monitoring devices.
Each monitor consists of a sensor inside containment, a remote analyzer and control panel located in the relay room, and a recorder in the control room.
The analyzers are capable of monitoring hydrogen content by volume from zero to ten percent over a containment atmosphere range of -2 to 60 psig, 40 to 290 degrees F,
and 0 to 100 percent relative humidity.
The inspector verified that maintenance and surveillance requirements stipulated in Technical Specifications are being properly adhered to.
The inspector determined that in May 1983 the hydrogen monitoring system supplier (Comsip, Inc.) issued a
CFR 21 notification identifying the potential for degraded performance of their analyzer cell standard configuration catalyst bed.
Comsip, Inc.
recommended replacement of the standard catalyst bed with a modified configuration catalyst bed to ensure optimum hydrogen monitoring system performance.
The licensee completed this modification to both monitoring trains on May 7, 1984.
On May 24, 1985 the analyzer cell for the 'A'rain hydrogen monitor was replaced due to leakage
identified during calibration.
No other significant hydrogen monitoring system problems have been observed since installation.
The inspector verified that Station Operations personnel perform an operability check, of the hydrogen monitors every shift (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />)
~
The monitors are normally placed in standby and operated continuously only during accident conditions.
The inspector verified that plant emergency procedures specify the hydrogen monitors are to be placed in operation in the event of a design basis accident.
IE Bulletin 79-06 79-06A and 79-06A Rev.
Res onses II.F. 1 These bulletins specified actions to be taken by the licensee to avoid occurrence of an event similar to that which occurred on March 28, 1979 at Three Mile Island, Unit No.
2 (TMI-2).
By letters dated April 28, 1979 and June 22, 1979 the licensee responded to these bulletin action items.
Crutchfield to White letter, dated May 28, 1980, issued an evaluation of the licensee's responses and concluded that the licensee had correctly interpreted the IE Bulletins and that all responses were acceptable with two minor exceptions.
The exceptions stipulated that final acceptability of two responses was pending verification of their implementation by the Office of Inspection and Enforcement.
The first item (response to Bulletin action item No. 7b)
requires the verification of the proper incorporation of approved Westinghouse generic Safety Injection (SI)
termination criteria into Station Emergency Procedures.
The inspector reviewed Emergency Procedures, E-l. 1, "Immediate Actions and Diagnostics for Spurious Actuations of SI, LOCA, Loss of Secondary Coolant, and Steam Generator Tube Rupture",
E-1
~ 2, "Loss of Reactor Coolant", E-1.3,
"Loss of Secondary Coolant" and E-1.4,
"Steam Generator Tube Rupture",
and verified Westinghouse SI termination criteria and pertinent precautions were specified.
The second item (response to Bulletin action item 7c) requires verification of the incorporation into Emergency Procedures the Westinghouse generic guidelines for reactor coolant pump tripping.
The inspector reviewed E-1. 1, E-l.2, E-l.3 and E-1.4 and verified that pump tripping criteria were specified.
Pump tripping criteria have received subsequent review and analysis by NRR since issuance of the licensee's May 28, 1980 evaluation.
Further analyses have been required of all licensees by the issuance of Generic Letters No.83-10c and d and No. 85-12.
Final NRR approval of the licensee's reactor
coolant pump tripping criteria and implementation review is being tracked under TMI Action plan item II.K.3.5.
Action plan item II.K.1 review is closed.
Reference:
EWR 2607C, Design Critera,
"Containment Hydrogen Monitoring", Revision 2, dated 11/5/81.
EWR 2607C, Safety Analysis,
"Containm'ent Hydrogen Monitoring", Revision 1, dated 11/5/81.
"Containment Hydrogen Monitoring System Calibration", Revision 4, dated 6/12/85.
PC-23.7,
"Containment Atmosphere Hydrogen Monitor",
Revision 5, dated 6/26/85.
0-9. 1, "Shift Relief Turnover-Auxiliary Operator",
Revision 8, dated 8/7/85.
SC-200,
"Emergency Response Organization/Responsibilities",
Revision 4, dated 8/15/84.
Comsip, Inc. to RG5E Corp letters dated May 10, 1983 and June 8, 1983.
M-72.3, "Hydrogen Analyzer Cell and/or Catalyst Bed Assembly Replacement Procedure",
Revision 2, dated 4/24/85.
U date of TMI Action Plan Items Remainin 0 en The inspector reviewed the following items to determine the progress being made by the licensee to complete the remaining TMI Action Plan commitments.
Short-Term Accident and Procedures Review I.C. 1 This action plan item requires the licensee to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures and conduct operator retraining.
By Kober to Zwolinski letter dated February 28, 1985, the licensee submitted the Procedures Generation Package,(PGP)
to NRR for review.
The PGP outlines the licensee's entire Emergency Procedures reanalyses, upgrade and training program.
Operator training and final validation of the revised Emergency Operating Procedures (EOPs)
commenced in July 1985.
Training is currently scheduled to be completed by September 1985 and the EOPs implemented by December 1985 '
Plant Safet Parameter Dis la Console I.D.2 This action plan item requires the licensee to install a
computer-based system which will provide operating personnel with the minimum set of parameters with which to define and monitor the safety status of,the plant.
During the upcoming 1986 Annual Refueling and Naintenance Outage the licensee plans to replace the plant process computer and install the new Safety Assessment System (SAS).
The SAS is not currently projected to be operational until 1987.
Instrumentation for Detection of Inade uate Core Coolin This action plan requires the licensee to provide additional instrumentation or'ontrols to supplement existing instrumen-tation in order to provide an unambiguous and easy-to-interpret indication of inadequate core cooling.
The installation of a reactor vessel level monitoring system is the one remaining requirement of this action plan item not completed, to date.
Engineering Work Request No.
2799 governs the installation of this station modification.
Work commenced on this project during the 1985 Outage and is scheduled to be completed during the 1986 Outage.
Automatic Tri of Reactor Coolant Pum s Durin Loss of Coolant Accident II.K.3.5 This action plan item requires the licensee to analyze and establish criteria for the continued operation or securing of reactor coolant pumps during various loss of coolant accidents.
Generic Letter No. 85-12, dated June 18, 1985, addresses the NRC staff's position regarding the Westinghouse Owners Group submittals on this action plan item and provides guidance concerning licensee implementation of the reactor coolant pump trip criteria.
The licensee's response to Generic Letter No. 85-12 is documented in Kober to Zwolinski letter, dated August 19, 1985.
The inspector will review the implementation of this item pending issuance of the Safety Evaluation Report by NRR.
U rade Emer enc Su ort Facilities III.A.1.2 This action plan item was clarified in Supplement 1 to NUREG-0737, dated December 17, 1982, and required numerous upgrades of the Emergency Support Facilities (Technical Support Center (TSC), Operational Support Center (OSC)
and Emergency Operations Facility (EOF)) to enhance communications, plant control, radiological protection, technical and operational support, emergency response and
accident recovery.
The installation of the Safety Assessment System (operational by 1987), with system accessibility in both the TSC and EOF, will complete the currently scheduled upgrading specified by this action plan item.
6.
Im rovin License Emer enc Pre aredness-Lon Term III.A.2 This action plan item requires the licensee to upgrade their emergency plans to provide reasonable assurance that adequate protective measures can be and will be taken in the event of a radiological emergency.
This item in conjunction with item III.A.1. 1, "Upgrade Emergency Preparedness" and item III.AD1.2,
"Upgrade Emergency Support Facilities", endorse the specific criteria of NUREG-0654 (FEMA-REP-1), "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparation in Support of Nuclear Power Plants",
and NUREG-0696,
"Functional Criteria for Emergency Response Facilities".
During the period of November 2-13, 1981 an NRC team conducted an appraisal (IR No. 50-244/81-22) of the state of Emergency Preparedness at Ginna station.
As a result, numerous items requiring resolution and improvement were identified to the licensee for action.
Subsequent inspections by the region-based inspectors (IR No.s 50-244/83-08, 83-25, 84-08 and 84-21)
have resolved the November 1981 appraisal items and identified additional items in the area of the meteorological program.
Final evaluation of the licensee's Meteorological Data Program (III.A.2.2) by the NRC staff is still pending.
Complete implementation of the licensee's Emergency Preparedness Program is also awaiting installation of the station's Safety Assessment System (SAS).
The SAS will provide a computerized information link between all the Emergency Support Facilities and the control room.
SAS is projected to be operational in 1987.
7.
Six additional items, not previously addressed, remain open pending final NRC staff evaluation.
Status of these items is being maintained by Region I or NRR staff.
These items are:
I.D. 1, Control Room Design Review II.D.3, Post Accident Sampling Capability II.F. 1, Attachment 1, Noble Gas Effluent Monitor II.F. 1, Attachment 2, Sampling and Analysis of Plant Effluents II.F. 1, Attachment 3, Containment High-Range Radiation Monitor III.D.3. 3, Improved Inp1 ant Iodine Instrumentation Under Accident Conditions
na
7.
Inte rated Plant Safet Assessment Review a.
The inspector reviewed the following item which was identified during the Systematic Evaluation Program (SEP).
The SEP reviews required equipment and/or procedural changes.
The topic number refers to a paragraph designation in NUREG-0821, "Integrated Plant Safety Assessment Final Report for the R.
E. Ginna Nuclear Power Plant",
December 1982.
b.
Electrical Penetrations of Reactor Containment 3.3.9 SEP topic VIII-4 reviewed electrical penetrations in the containment structure to ensure that they do not fail from electrical faults during a high energy line break.
The NRC staff conducted an audit comparing sample containment electrical penetrations with current licensing criteria for protection against fault and overload currents following a postulated accident.
The staff concluded that for a loss-of-coolant-accident (LOCA) environment inside containment, the backup protection for some penetrations did not conform to current licensing criteria.
By Maier to Crutchfield letter, dated June 9,
1981, the licensee committed to provide the necessary circuit pro-tection to ensure that electrical penetrations conform to current licensing criteria.
The licensee proposed circuit protection upgrades, to ensure electrical penetrations conform to current licensing criteria, were reviewed and found acceptable by the NRC staff as documented in NUREG-0944, "Safety Evaluation Report related to the full-term operating license for R.
E. Ginna Nuclear Power Plant", dated October 1983.
The inspector reviewed the below listed station modification packages to verify satisfactory completion of the electrical penetration upgrades.
The modification packages were reviewed and discussed with the licensee to verify:
procedural compliance; administrative completeness; proper station review and acceptance; and satisfactory resolution of surveillance and nonconformance reports.
No discrepancies were noted.
Station Modification Procedure, (SM)-2929.2,
"Relay Modification Electrical Penetrations Hardware Installation",
Revision 0, dated 3/13/82, completed 12/19/83.
SM-2929.8,
"Termination and Splicing of Cables for Auxiliary Breaker Cabinet MCC 1C", Revision 1, dated 5/16/83, completed 11/28/83.
SM-2929.9,
"Termination and Splicing of Cables for Auxiliary Breaker Cabinet MCC 1D, Revision 0, dated 5/2/83, completed 11/28/8 SM-2929.10, "Installation and Testing of the 1A Control Rod Shroud Fan Auxiliary Breaker",
Revision 0, dated 5/19/83, completed 9/28/84.
SM-2929. 11, "Installation and Testing of the 1B Control Rod Shroud Fan Auxiliary Breaker",
Revision 0, dated 5/19/83, completed 6/23/83.'M-2929.
12, "Termination and Testing of Auxiliary Breaker for Lighting Transformer 1D", Revision 0, dated 6/24/83, completed 10/12/83.
8.
S ent Fuel Cask Unloadin On July 1, 1985, the licensee received the first of eighty-one (81) spent fuel assemblies scheduled to be returned from the U.S. Department of Energy, West Valley Demonstration Project in West Valley, New York.
The inspector observed the removal of the first spent fuel assembly from its shipping cask and placement into the Spent Fuel Pool on August 8, 1985.
The inspector verified the initial conditions were satisfied and reviewed the step-by-step sequence of events specified in the controlling procedure RF-8.7,
"NLI 1/2 Spent Fuel Shipping Cask Handling and Unloading".
No discrepancies were noted.
The inspector observed that radiological control methods were conservative and practicable, equipment handling was professional and well supervised, and that Quality Control (QC) practices and QC personnel involvement were satisfactory.
9.
IE Bulletin Follow-u The inspector reviewed licensee actions on the following IE Bulletin to determine that the written response was submitted within the required time period, that the response included the information required including adequate corrective action commitments, and that licensee management provided adequate dissemination of the bulletin and the response.
The review included discussions with licensee personnel and observations of the item discussed below.
IE Bulletins 79-06,79-06A and 79-06A Rev.
1:
Review of Operational Errors and System Misalignments identified during the Three Mile Island Incident.
All corrective actions taken in response to these bulletins have been satisfactorily completed and reviewed or tracked via TMI Action Plan item listings, (see paragraph 6.b.3 for details).
These bulletins are administratively close.
Review of Periodic and S ecial Re orts Upon receipt, periodic and special reports, submitted by the licensee pursuant to Technical Specification 6.9. 1 and 6.9.3 were reviewed by the inspector.
This review included the following considerations:
the reports contained the information required to be reported by NRC require-ments; test results and/or supporting information were consistent with design predictions and performance specifications; and the validity of the reported information.
Within the scope of the'bove, the following report was reviewed by the inspector:
Monthly Operating Report for July 1985.
11.
Exit Interview At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection, scope and findings.
Based on the NRC Region I review of this report and discussion held with licensee representatives on September 13, 1985, it was determined that this report does not contain information subject to
CFR 2.790 restriction Item Number I.A.1.1-I.AF 1.2 I.A.1 '
I.A.2.1 I.C.2 I.C.3 I.C.4 I.C.5 I.C.6.
II.B.1 II.B.P.
II.B.4 II.D.3 II.E.F 1'I.E.1.2 II.E.3.1 II.E.4.1 II.ED 4.2 II.F.1.4 II.G.1 II. K. 3'
I.K.3,9 II.K.3.10 ATTACHMENT A CLOSED TMI ACTION ITEMS REVIEWED Titl e Shift Technical Advisor Shift Supervisor Responsibilities Shift Manning Immediate Upgrading of RO and SRO Training Qualifications Shift & Relief Turnover Procedures Shift Supervisor Responsibility Control Room Access Feedback of Operating Experience Verify Correct Performance of Operating Activities Reactor Coolant System Vents Plant Shielding Training for Mitigating Core Damage Valve Position Indication Auxiliary Feedwater System" Evaluation Auxiliary Feedwater System Initiation & Flow Emergency Power for Pressurizer Heaters Dedicated Hydrogen Penetrations Containment Isolation Dependability Accident Monitoring - Containment Pressure Power Supplies for Pressurizer Rel'ief Valves, Block Valves
& Level Indicators Final Recommendations, B&0 Task Force PID Controller Proposed Anticipating Trip Modification
Attachment A
II.K.3~ 12.B II.K.3.25.B.2 Anticipating Reactor Trip or Turbine Trip Power on Pump Seals Modification III.A.1.2.1 III.0.1.1 111.0.3.4 Upgrade Emergency Support Faci 1 ities (Interim)
Primary Coolant Outside Containment Contr ol Room Habi tabi 1 ity